ML18036A763

Proposed TS Table 4.2.A, Surveillance Requirements for Primary Containment & Reactor Bldg Isolation Instrumentation & Table 4.2.B, Surveillance Requirements for Instrumentation That Initiates or Controls Cscs.

ML18036A763
Person / Time
Site:	Browns Ferry
Issue date:	07/02/1992
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18036A762	List: ML18036A761 ML18036A763
References
TVA-BFN-TS-314, NUDOCS 9207130077
Download: ML18036A763 (25)

Text

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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNIES FERRY NUCLEAR PLANT (TVA BFN TS-314) 9207f30077 920702 ~(

PDR ADClCK 0~00 P PDR

E II

UNIT 1 EFFECTIVE PAGE LIST TECHNICAL SPECIFICATION 314 REMOVE INSERT 3.2/4.2-16 3.2/4.2-16 3.2/4.2-40 3.2/4.2-40 3.2/4.2-45 3.2/4.2-45 3.2/4.2-61 3.2/4.2-61

TABLE 3.2.B (Continued)

Hinimum No.

Operable Per

~Tri ~1 Fn in Tri L v 1 in ~AIhh R mark Instrument Channel 450 psig + 15 l. Below trip setting permissive Reactor,Low Pressure for opening CSS and LPCI (PS-3-74 A & B, SW ¹2) admission valves.

(PS-68-95, SW ¹2)

(PS-68-96, SW ¹2)

Instrument Channel 230 psig + 15 1. Recirculation discharge valve Reactor Low Pressure actuation.

(PS-3-74 A & 8, SW ¹1)

(PS-68-95, SW ¹1)

(PS-68-96, SW ¹1)

Core Spray Auto Sequencing 6< t <8 sec. 1. With diesel power Timers (5) 2. One per motor LPCI Auto Sequencing 0< t <1 sec. 1. With diesel power Timers (5) 2. One per motor RHRSW Al, B3, Cl, and D3 13< t <15 sec. 1. With diesel power Timers 2. One per pump Core Spray and LPCI Auto 0< t <1 sec. 1. With normal power Sequencing Timers (6) 6< t <8 sec. 2. One per CSS motor 12< t <16 sec. 3. Two per RHR motor 18< t < 24 sec.

RHRSW Al, B3, Cl, and D3 27< t < 29 sec. 1. With normal power Timers 2. One per pump

TABLE 4.2.A SURVEILLANCE REgUIREHENTS FOR PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION

~Functi n Func ion 1 T libra i n Fr u nc Ins rum nt Ch k Instrument Channel (5) once/day Reactor Low Water Level (LIS-3-203A-D, SW 2-3)

Instrument Channel-- (31) once/18 months None Reactor High Pressure (PS-68-93 & -94)

Instrument Channel- once/3 month once/day Reactor Low Mater Level (LIS-3-56A-D, SW 01)

Instrument Channel- (5) N/A High Drywell Pressure (PS-64-56A-D)

Instrument Channel- once/3 months (29) (5) once/day High Radiation Hain Steam Line Tunnel 4

Instrument Channel- once/3 months (27) (29) once/operating cycle (28) None Low Pressure Hain Steam I Line (PT-1-72, -76, -82, -86)

C)

Instrument Channel once/3 months (27) (29) once/operating cycle (28) once/day High Flow Hain Steam Line (dPT-1-13A-D, -25A-D, -36A-D, -50A-0)

TABLE 4.2.B (Continued)

SURVEILLANCE REgUIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fn in Fn inl 7 C libra i n In rum nt he k Core Spray Auto Sequencing Timers (4) once/operating cycl e none (Normal Power)

Core Spray Auto Sequencing Timers (4) once/operating cycle none (Diesel Power)

LPCI Auto Sequencing Timers (4) once/operating cycle none (Normal Power)

LPCI Auto Sequencing Timers (4) once/operating cycle none (Diesel Power)

RHRSW Al, B3, Cl, D3 Timers (4) once/operating cycle none (Normal Power)

RHRSW Al, B3, Cl, 03 Timers (4) once/operating cycle none (Diesel Power)

ADS Timer (4) once/operating cycle none

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NOTES FOR TABLES 4 2 A THROUGH 4 2 L exce t 4 2.D AND 4 2 K (Cont'd) 2'6. This instrument check consists of comparing the background si'gnal levels for all valves for consistency and for nominal expected values (not required during refueling outages).

27. Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.

28. Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment, of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

29. The functional test frequency decreased to once/3 months to reduce challenges to relief valves per NUREG-0737, Item II.K.3.16.

30. Calibration shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one-point source check of the detector below 10 R/hr with an installed or portable gamma source.

31. Functional tests shall be performed once/3 months.

BFN 3.2/4.2-61 Unit 1

UNIT 3, EFFECTIVE PAGE LIST TECHNICAL SPECIFICATION .314 REMOVE INSERT 3.2/4.2-16 3.2/4.2-16 3.2/4.2-39 3.2/4.2-39 3.2/4.2-44 3.2/4.2-44 3.2/4.2-60 3.2/4.2-60

TABLE 3.2.B (Continued)

Hinimum No.

Operable Per

~Tri y,LL11 Fn in Tri L v 1 tin ~Aion R ark lA 2 Instrument Channel 450 psig + 15 1. Below trip setting permissive Reactor Low Pressure for opening CSS and LPCI (PS-3-74 A L B, SW ¹2) admission valves.

(PS<8-95, SW ¹2)

(PS-68-96, SW ¹2)

Instrument Channel 230 psig + 15 1. Recirculation discharge valve Reactor Low Pressure actuation.

(PS-3-74 A 5 B, SW ¹1)

(PS-68-95, SW ¹1)

(PS-68-96, SW ¹1)

Core Spray Auto Sequencing 6< t <8 sec. 1. With diesel power Timers (5) 2. One per motor LPCI Auto Sequencing 0< t <1 sec. 1. With diesel power Timers (5) 2. One per motor 4J RHRSW A3, Bl, C3, and 01 .13< t <15 sec. l. With diesel power Timers 2. One per pump C Core Spray and LPCI Auto 0< t <1 sec. 1. With normal power Sequencing Timers (6) 6< t <8 sec. 2. One per CSS motor I 12< t <16 sec. 3. Two per RHR motor 18< t < 24 sec.

RHRSW A3, Bl, C3, and 01 27< t < 29 sec. 1. With normal power Timers 2. One per pump

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TABLE 4.2.A SURVEILLANCE REgUIREHENTS FOR PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION

~Fn ~in Fn inlT libr ti n Fr n Instrument Channel (5) once/day Reactor Low Water Level (LIS-3-203A-D, SW 2-3)

Instrument Channel (30) once/18 months None Reactor High Pressure (PS-68-93 & -94)

Instrument Channel once/3 month once/day Reactor Low Mater Level (LIS-3-56A-D, SW 01)

Instrument Channel- (5) N/A High Drywell Pressure (PS-64-56A-D)

Instrument Channel once/3 months (27) (5) once/day High Radiation Hain Steam Line Tunnel Instrument Channel- once/3 months (27) once/3 months None Low Pressure Hain Steam Line Instrument Channel- once/3 months (27) once/3 months once/day High Flow Hain Steam Line

TABLE 4.2.8 (Cont'd)

SURVEILlANCE RE(lUIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fn in Fun i nl T t alibr ti n Core Spray Auto Sequencing Timers (4) once/operating cycle none (Normal Power)

Core Spray Auto Sequencing Timers (4) once/operating cycle none (Diesel Power)

LPCI Auto Sequencing Timers (4) once/operating cycle none (Normal Power)

LPCI Auto Sequencing Timers (4) once/operating cycle none (Diesel Power)

RHRQl A3, Bl, C3, Dl Timers (4) once/operating cycle none (Normal Power)

RHRSW A3, Bl, C3, Dl Timers (4) once/operating cycle none (Diesel Power)

ADS Timer (4) once/operating cycle none

O.

NOTES FOR TABLES 4 2 A THROUGH 4 2 L exce t 4.2 D AN 4 2 K (Continued)

26. This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outages).

27. Functional test frequency decreased to once/3 months to reduce the challenges to relief valves per NUREG-0737, Item II.K.3.16.

28. Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.

29. Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so its output relay changes state at or more conservatively than the analog equivalent of the trip level settings.

30. Functional tests shall be performed once/3 months.

BFN 3.2/4.2-60

-Unit 3

ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT DESCRIPT10N AND JUSTIFICATION FOR THE PROPOSED CHANGE Summar of Chan e Unit 1

1. Revise Table 3.2.B to delete the function "Instrument Channel Reactor Low Pressure (PS-68-93 & 94, SW gl)".

2. Revise Table 4.2.A, function "Instrument Channel Reactor High Pressure" as follows:

a. Add pressure switch numbers (PS-68-93 & 94).

b. Change functional test note to (31) from present note (1) ~

c. Change calibration frequency to once/18 months from present once/3 months.

3. Delete the function "Instrument Channel Reactor Low Pressure, (PS-68-93 & 94)" from Table 4.2.B.

4. Revise notes for Table 4.2.A to add note 31, as follows:

"31. Functional tests shall be performed once/3 months."

Unit 3

1. Revise Table 3.2.B to delete the function "Instrument Channel Reactor Low Pressure (PS-68-93 & 94, SW N1)".

2. Revise Table 4.2.A, function "Instrument Channel Reactor High Pressure" as follows:

a. Add pressure switch numbers (PS-68-93 & 94).

b. Change functional test note to (30) from present note (1) ~

c. Change calibration frequency to once/18 months from present once/3 months.

3. Delete the function "Instrument Channel Reactor Low Pressure, (PS-68-93 & 94)" from Table 4.2.B.

4. Revise notes for Table 4.2.A to add note 30, as follows:

"30. Functional tests shall be performed once/3 months."

ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT DESCRXPTXON AND iTUSTXFXCATXON FOR THE PROPOSED CHANGE Reason for Chan e Non-class 1E reactor pressure switches PS-68-93 and PS-68-94 with two sets of contacts (microswitches), are being replaced with Class 1E Static-0-Ring (SOR) pressure switches with one set of contacts. The present switches have an adjustable setpoint, range from 50 to 1200 psig. The bistable input contacts open and close at a setpoint. of 100 plus or minus 15 psig. As a result of the wide range of the switches, excessive drift has caused unacceptable instrument accuracy in the lower pressure ranges.

The replacement pressure switches have a setpoint range of 20 to 180 psig and an accuracy of 1.0 percent, of upper range limit.

However, the Class 1E pressure switches can only be purchased with one set of contacts. Therefore, to allow installation of the SOR pressure switches, the isolation logic for two RHR isolation valves will be modified to eliminate a redundant permissive signal generated by one set of contacts. The proposed change deletes this function from Technical Specifications. The proposed change also includes new calibration and surveillance intervals appropriate for the new reactor pressure switches.

Justification for Chan e This technical specification change affects the shutdown cooling portion of the Residual Heat Removal (RHR) System. The current pressure switches, PS-68-93 and PS-68-94, each contain two microswitches (SW gl, SW 42). TVA proposes to delete SW N1 from isolation logic since its function is redundant to the limit switches of Residual Heat Removal (RHR) inboard/outboard isolation valves FCV-74-47 and FCV-74-48. SW Nl closes on decreasing reactor pressure at a setpoint of 100 plus or minus 15 psig and provides a low pressure permissive signal to the isolation logic for RHR valves FCV-74-53 and FCV-74-67. A permissive signal is also provided to these valves whenever valves FCV-74-47 and FCV-74-48 are in the open position. Valves FCV-74-47 and FCV-74-48 cannot be opened unless reactor pressure is below 100 psig. SW 42 closes on increasing reactor pressure at a setpoint of 100 plus or minus 15 psig and generates a signal to automatically close the inboard/outboard RHR isolation valves FCV-74-47 and FCV-74-48 when the reactor pressure approaches the shutdown cooling piping design pressure. The maximum allowable pressure is 133 psig. The function of this signal is to protect the shutdown cooling system piping and components from overpressurization.

A low pressure signal generated from either pressure switch (PS-68-93 or PS-68-94) in conjunction with not fully closed signals from FCV-74-47 and FCV-74-48, and the absence of a containment

pit r ~ lh l 1 +

ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGE Justification for Chan e (continued) isolation signal satisfies the opening logic for valves FCV-74-53 and FCV-74-67. The redundant low pressure permissive will be eliminated since SW gl will be removed and replaced by a jumper which will in effect provide a permanent low pressure permissive.

SW g1 provides a non-safety permissive signal for isolation of isolation valves FCV-74-53 and FCV-74-67. This permissive signal is redundant to the permissive signal provided by the limit switches on FCV-74-47 and FCV-74-48 when these valves are open.

Since valves FCV-74-47 and FCV-74-48 cannot be opened unless reactor pressure is below 100 psig, the permissive provided by switch SW g1 is unnecessary.

The increased functional test and calibration test intervals are justified by the improved reliability of the Class 1E SOR pressure switches. The existing pressure switches are not class 1E and have been subject to problems with accuracy and drift. As a result, they are being replaced with the more reliable and accurate SOR pressure switches. Although the pressure switches are not required to operate during a 10 CFR 50.49 event, the purchased pressure switches have been subjected, by the vendor, to mechanical aging by cycling the units 33,000 times at the upper adjustable limit and exposing the unit to the LOCA environment to determine the effect on switch performance. The vendor has successfully demonstrated the switch capability to withstand the test conditions. The switches will be used to isolate the RHR shutdown cooling suction valves when the reactor vessel pressure is above the shutdown cooling range. The anticipated duty cycle would be based on quarterly functional testing and the number of shutdowns. 'TVA considers a duty cycle of ten per year as a conservative number.

TVA has established, by calculation, setpoint values based on a 22 1/2 month calibration interval. Drift due to pressurization of the switches above the upper adjustable limit is considered in the calculation. These values have been determined by TVA calculations for Unit 2. The proposed values for Units 1 and -3 are based on the Unit 2 calculation and will be confirmed by unit speci'fic calculation's prior to use.

The proposed change to the functional test interval from once each month to once every three months is not expected to have. an adverse impact on plant safety since the improved reliability of the new pressure switches provides substantial assurance that the switches will properly perform their function.

p:t ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGE Justification for Chan e (continued)

The proposed change to the calibration test interval from once every three months to once every 18 months is supported by the accuracy and drift characteristics of the class 1E pressure switches. A setpoint and scaling calculation has been performed that demonstrates the acceptability of the increased interval.

The calculation determined the effect of a 22-1/2 month (18 month plus the allowable 25%) interval between calibrations using vendor supplied instrument accuracy and drift information. The results of the calculation demonstrated that the increased calibration interval does not adversely affect the function of the instruments.

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ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN)

PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSXDERATXON Descri tion of Pro osed Technical S ecification Amendment Non-class 1E reactor pressure switches PS-68-93 and PS-68-94 with two sets of contacts (microswitches), are being replaced with more accurate Class 1E Static-0-Ring (SOR) pressure switches with one set of contacts. The present switches have an adjustable set point. range from 50 to 1200 psig. The bistable input contacts open and close at a setpoint of 100 plus or minus 15 psig. As a result of the wide range of the switches, excessive drift has caused unacceptable instrument accuracy in the lower pressure ranges. The Class lE SOR pressure switches have a setpoint range of 20 to 180 psig and an accuracy of 1.0 percent of upper range limit. However, the Class 1E pressure switches can only be purchased with one set of contacts. Therefore, to allow installation of the SOR pressure switches, the isolation logic for two RHR isolation valves will be modified to eliminate an unnecessary permissive signal generated by one set of contacts.

The proposed change deletes this function from Technical Specifications. The proposed change also includes new calibration and surveillance intervals appropriate for the new reactor pressure switches.

Basis for Pro osed No Si nificant Hazards Consideration Determination The NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license involves no significant hazards consideration in accordance with the proposed if amendment operation of the facility would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

1. The proposed change does not significantly increase the probability or consequences of an accident previously evaluated.

The existing non-class 1E pressure switches (PS-68-93 and 94) are being replaced by class 1E pressure switches to resolve problems with inadequate pressure switch accuracy and excessive drift. The existing pressure switches contain two internal microswitches (SWN1, SWN2) while the replacement switches contain one internal microswitch. As a

0 ENCLOSURE 3 BROWNS FERRY NUCLEAR PLANT (BFN)

PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Basis for Pro osed No Si nificant Hazards Consideration Determination (continued) result, the function of present SW gl, which is to provide a low pressure permissive signal to the isolation logic for RHR valves FCV-74-53 and FCV-74-67, is being deleted from Tables 3.2.B and 4.2.B by this change. This function is redundant to the limit switches on RHR valves FCV-74-47 and FCV-74-48. As such, it analysis.

is not required nor was Changes are also being it considered in the FSAR made to Table 4.2.A to reflect the revised functional testing and calibration requirements for the new pressure switches.

No new failure modes have been identified for the proposed changes. Misoperation of the replacement pressure switches could not increase the probability or consequences of any accident previously evaluated in the plant Final Safety Analysis Report (FSAR). Further, the replacement pressure switches do not require relocation, do not adversely affect system function or operations, and do not, adversely affect other systems or components. Therefore, this change will not significantly increase the probability or consequences of any accident previously evaluated in the FSAR.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The function and operation of the affected systems are not changed by the amendment. Seismic qualification of the affected components remain intact after this modification and other systems will not be adversely affected. Operation and failure modes of the replacement switches can cause no different effects than the existing switches. Thus, the credible failure modes of the replacement switches would be bounded by existing FSAR Section 14.6.3.3.2 accident analysis.

Therefore, this modification will not create the possibility of an accident of a different type than any previously evaluated in the FSAR.

3. The proposed change does not involve a significant reduction in a margin of safety.

The change replaces the existing non-Class 1E pressure switches with Class 1E pressure switches that are more

ENCLOSURE 3 3 BROWNS FERRY NUCLEAR PLANT (BFN)

PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION for Pro osed No Si nificant Hazards Consideration

'asis Determination (continued) accurate. In addition, one of the two contacts from each pressure switch will be removed from the current valve control logic. This contact is redundant to other logic that controls these valves and is not required for proper operation of any logic recpxired for Technical Specification compliance.

The margin of safety defined by the bases for Technical Specifications 3.2.A/4.2.A (Primary Containment and Reactor Building Isolation Functions) and 3.2.B/4.2.B (Core and Containment Cooling Initiation & Control) is not reduced by this modification. This modification results in increased instrument accuracy and a reduction of failure modes as a result of the deletion of redundant contacts. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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